• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.789-794
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• 1998

중대사고 해석 전산 코드 국산화의 필요성이 대두되고 있는 이때 우리가 개발해야 할 코드의 요건을 다음 여섯 가지로 정리하였다: 1) 종합적인 해석 코드. 2) 1차계통과 격납건물 모사 능력의 유연성, 3) 자세한 발전소 거동 모사, 4) 사용자 편의성, 5) 개선 및 새로운 모델 접목의 용이성. 그리소 6) 최신 모델 포함. 이런 관점에서 기존의 중대사고 해석코드를 분석한 결과 코드 개발의 기준 코드로 MELCOR를 선정하였다. MELCOR는 계통 모사의 유연성 때문에 상용 발전소 뿐만 아니라 앞으로 개발 계획 중인 차세대나 중소형 원자로까지도 확장이 가능하며, 상세한 열수력 기본 지배 방정식을 활용하고. 모델 분석 및 개선에 필요한 코드에의 자유로운 접근이 허용되며, 지속적인 코드 개선이 이루어져 최신 모델을 보유하고 있다. 이미 MELCOR는 상당한 수준의 결과를 예측하고 있기만. 노심 손상 모델을 개선하고 격납건물 안에서의 주요 현상 모사 모델을 추가하며. 또한 국내에서 이루어지고 있는 SONATA 실험이나 증기 폭발 실험 결과들을 MELCOR에 반영하는 것이 가급적 짧은 시간에 기술 자립을 이를 수 있는 방법으로 판단된다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.20 no.2
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• pp.184-192
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• 2024

This paper evaluates the ultimate pressure capacity (UPC) of a prestressed concrete containment building, considering the loss of prestress. A three-dimensional finite element model with openings was utilized for the UPC evaluation, and material properties reflecting severe accident temperatures were applied. The failure criteria were established based on the strain of the liner plate in accordance with the regulatory guidelines. When the prestress was reduced by approximately 14% from its initial value, the UPC decreased by about 4%. This study acknowledges limitations in the determination of nonlinear material properties, realistic prestress losses, and failure criteria. Further research to address these limitations could enhance the accuracy of UPC evaluation.

• Journal of the Earthquake Engineering Society of Korea
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• v.12 no.6
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• pp.55-63
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• 2008

This paper presents a seismic analysis technique for a 3D soil-structure interaction system in a frequency domain, based on the finite element formulation incorporating frequency-dependent infinite elements for the far field soil region. Earthquake input motions are regarded as traveling P, SV and SH waves which are incident vertically from the far-field soil region, and then equivalent earthquake forces are calculated using impedances of infinite soil by dynamic infinite elements and traction and displacement from free field response analysis. For verification and application, seismic response analyses are carried out for a multi-layered soil medium without structure and a typical nuclear power plant in consideration of soil-structure interaction. The results are compared with the free field response using a one-dimensional analytic solution, and a dynamic response of an example structure from another SSI package.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.231-236
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• 2011

When the Fukushima nuclear power plant accident occurred in March of 2011, a hydrogen explosion in the reactor building at the 4th unit of Fukushima plants led to a big surprise because the full core of the unit 4 reactor had been moved and stored underwater at the spent nuclear fuel storage pool for periodic maintenance. It was because the possible criticality in the fuel storage pool by coolant loss may yield more severe situation than the similar accident happened inside the reactor vessel. Fortunately, it was assured to be evitable to an anxious situation by a look of water filled in the storage pool later. In the paper, the safety state of the spent fuel storage pool and rack structures of the domestic nuclear plants would be roughly reviewed and compared with the Fukushima plant case by engineering viewpoint of potential severe accidents.

• Fire Science and Engineering
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• v.32 no.1
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• pp.57-65
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• 2018

In this study, a post-fire safe shutdown analysis (PFSSA) including multiple spurious operation (MSO) treatments for cables was conducted with an imaginary nuclear power plant for training using a deterministic fire analysis code. The imaginary nuclear power plant for the training consisted of a reactor containment building and an auxiliary building, including a total of 22 fire areas. The equipment including valves, pumps, emergency diesel generators, switch gears, motor control centers, and logic controllers were located in each fire area of the imaginary plant. It was assumed that each equipment is connected with two cables and that each cable passes through the fire areas along the cable trays. A database containing the information on the equipment and cables for the imaginary plant was constructed for the fire area analysis. The fire area analysis was performed for several assumed MSO scenarios, equipment logics, and cable logics. A mitigation measure using a three hour rated wrap was applied to the failed cables and cable trays after the fire area analysis.

• Magazine of the Korean Society of Agricultural Engineers
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• v.38 no.3
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• pp.82-91
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• 1996

This is a basic study for the static and dynamic analysis on the elasto-plastic and elasto-viscoplastic of an axi-symmetric shell. The objective of this study was to investigate the mechanical characteristics of a nuclear reinforced concrete containment structure, which was selected as a model, by a numerical analysis using a finite element method. The structure was modeled with discrete ring elements of 8-noded isoparametric element rotating against the symmetrical axis, and the interaction between the foundation and the structure was modeled by Winkler's model. Also, the meridional tendon was modeled with 2-node truss elements, and the hoop tendon was done with point elements in two degrees of freedom. The effect of the tendon was considered without the increasement in total degree of freedom as the stiffness matrix of modeled tendon elements was assembled on the stiffness matrix of ring elements linked with the tendon. The results obtained from the analysis of an example were summarized as follows : 1. The stresses in the hoop direction on the interior and exterior surfaces of the structure were shown in changes of similar trend, and high stresses appeared on the structure wall 2. The stresses in the meridional direction on the interior and exterior surfaces were shown in change of different trend. Especially, the stresses at the junctions between the dome and the wall and between the wall and the bottom plate of the structure were very high, compared with those at other parts of the structure. 3. The stress changes in the direction of thickness on the crown of the dome were much linearly distributed. However, as the amount of tendon increased, the stresses in the upper and lower parts of the wall established with the tendon were shown stress concentration. 4. The stress changes in the direction of thickness on the center of the structure wall was linearly distributed in the all cases, and special stress due to the use of the tendon was not shown.

• Journal of the Korean Society of Safety
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• v.30 no.5
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• pp.123-130
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• 2015

Wolsong Unit 2,3&4 has been performing a containment integrity test during power operation. This test could impact to the safe operation during test. If an accident occurs during pressure dropping phase, reactor trip can be delayed because of the increased pressure difference which causes a time delay to reach the trip set-point. On the contrary, if an accident occurs during pressure increasing phase, reactor trip could be accelerated because the pressure difference to the trip set-point decrease. Point Lepreau nuclear power plant, which installed GCLMS (Gross Containment Leakage Monitoring System) in 1990, has discontinued the test since 1992 due to these adverse effects. Therefore, we evaluated the risk to obviate the GCLMS test based on PWR's ILRT (Integrated Leak Rate Test) extension methodologies. The results demonstrate that risk increase rate is not high in case of performing only ILRT test at every 5 years instead of doing GCLMS test at every 1.5 years. In addition, the result shows that GCLMS test can be removed on a risk-informed perspective since risk increasement is in acceptable area of regulatory acceptance criteria.

• Journal of the Korean Society of Safety
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• v.33 no.3
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• pp.92-98
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• 2018

The flow of cooling water in a passive containment cooling system (PCCS), used to remove heat released in design basis accidents from a concrete containment of light water nuclear power plant, was conducted in order to investigate the thermo-fluid equilibrium among many parallel tubes of PCCS. Numerical simulations of the subcooled boiling flow within a coolant loop of a PCCS, which will be installed in innovative pressurized-water reactor (PWR), were conducted using the commercially available computational fluid dynamics (CFD) software ANSYS-CFX. Shear stress transport (SST) and the RPI model were used for turbulence closure and subcooled flow boiling, respectively. As the first step, the simplified geometry of PCCS with 36 tubes was modeled in order to reduce computational resource. Even and uneven thermal loading conditions were applied at the outer walls of parallel tubes for the simulation of the coolant flow in the PCCS at the initial phase of accident. It was observed that the natural circulation maintained in single-phase for all even and uneven thermal loading cases. For uneven thermal loading cases, coolant velocity in each tube were increased according to the applied heat flux. However, the flows were mixed well in the header and natural circulation of the whole cooling loop was not affected by uneven thermal loading significantly.

• 한국전산유체공학회:학술대회논문집
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• 2009.11a
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• pp.121-128
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• 2009

During a hypothetical high-pressure accident in a nuclear power plant (NPP), molten corium can be ejected through a breach of a reactor pressure vessel (RPV) and dispersed by a following jet of a high-pressure steam in the RPV. The dispersed corium is fragmented into smaller droplets in a reactor cavity of the NPP by the steam jet and released into other compartments of the NPP by a overpressure in the cavity. The fragments of the corium transfer thermal energy to the ambient air in the containment or interact chemically with steam and generate hydrogen which may be burnt in the containment. The thermal loads from the ejected molten corium on the containment which is called direct containment heating (DCH) can threaten the integrity of the containment. DCH in a NPP containment is related to many physical phenomena such as multi-phase hydrodynamics, thermodynamics and chemical process. In the evaluation of the DCH load, the melt dispersion rates depending on the RPV pressure are the most important parameter. Mostly, DCH was evaluated by using lumped-analysis codes with some correlations obtained from experiments for the dispersion rates. In this study, MC3D code was used to evaluate the dispersion rates in the APR1400 NPP during the high-pressure accidents. MC3D is a two-phase analysis code based on Eulerian four-fields for melt jet, melt droplets, gas and water. The dispersion rates of the corium melt depending on the RPV pressure were obtained from the MC3D analyses and the values specific to the APR1400 cavity geometry were compared to a currently available correlation.

• Proceedings of the Korea Concrete Institute Conference
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• 1993.04a
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• pp.219-224
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• 1993

본 연구는 철근콘크리트 격납구조물에서 가상의 냉각제 유출사고에 의한 온도하중과 압력에 따른 거동을 알아보기 위한 비선형 해석을 수행하였다. 시간에 따른 온도하중을 결정하기 위하여 과도온도해석을 통해 격납구조물 단면내의 온도분포를 구하였다. 구조물은 기하학적 비선형성과 재료비선형성을 고려한 판 및 쉘요소로 이상화되며, 쉘요소는 두께방향에 따라 변하는 응력을 고려하기 위해 몇 개의 층으로 이루어진 모델을 사용하였다. 열응력은 인접한 두시간 단계에서의 온도차를 하나의 하중증가로 고려하여 초기변형 문제로 변환하여 결정되었다. 본연구에서의 수치실험에 의하여 과도온도해석에 근거한 비선형온도경사를 고려할 때의 변위가 고려하지 않을 때의 변위에 비해 크게 나타남을 알 수 있다.